NL-17-115, Response to Request for Additional Information Re Inter-Unit Transfer of Spent Fuel, Attachments 1, 2, 3 (non-proprietary Version of Report HI-2094289, Rev. 9) and 7 Enclosed

Response to Request for Additional Information Re Inter-Unit Transfer of Spent Fuel, Attachments 1, 2, 3 (non-proprietary Version of Report HI-2094289, Rev. 9) and 7 Enclosed

ML17289A653
Person / Time
Site:	Indian Point
Issue date:	10/02/2017
From:	Vitale A Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML17289A619	List: ML17289A655 NL-17-115, Response to Request for Additional Information Re Inter-Unit Transfer of Spent Fuel, Attachments 1, 2, 3 (non-proprietary Version of Report HI-2094289, Rev. 9) and 7 Enclosed
References
CAC MF8991, CAC MF8992, NL-17-115
Download: ML17289A653 (53)

Text

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Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box249 Buchanan, NY 10511-0249 Tel (914) 254 6700 Anthony J Vitale Site Vice President NL-17-115 October 2, 2017 U.S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738

SUBJECT:

Response to Request for Additional Information Regarding Inter-Unit Transfer of Spent Fuel Indian Point Nuclear Generating Unit Nos. 2 and 3 (CAC Nos. MF8991 and MF8992)

Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64

REFERENCES:

1) Entergy letter dated December 14, 2016, "Indian Point Nuclear Power Plant Units 2 and 3 Proposed License Amendment Regarding the Inter-Unit Transfer of Spent Fuel," (NL-16-118) (ML16355A067)

2) NRC letter dated April 11, 2017, "Indian Point Nuclear Generating Unit Nos. 2 and 3 - Supplemental Information Needed for Acceptance of Requested Licensing Action RE: Amendment of Inter-Unit Transfer of Spent Fuel," (CAC Nos. MF8991 and MF8992) (ML17100A128)

3) Entergy letter dated April19, 20.17, "Response to Request for Supplemental Information Needed for Acceptance of Requested Licensing Action Regarding Amendment of Inter-Unit Transfer of Spent Fuel," (NL-17-044) (ML17114A467)

4) NRC letter dated April 26, 2017, "Indian Point Nuclear Generating Unit Nos. 2 and 3 - Acceptance of requested Licensing Action RE:

Amendment of Inter-Unit Transfer of Spent Fuel," (CAC Nos. MF8991 and MF8992) (ML17115A048)

5) NRC letter dated July 17, 2017, "Indian Point Nuclear Generating Unit Nos. 2 and 3 - Request for Additional Information Regarding Inter-Unit Transfer of Spent Fuel," (CAC Nos. MF8991 and MF8992)

(ML 17197AOOO)

6) NRC letter dated August 10, 2017, "Indian Point Nuclear Generating Unit Nos. 2 and 3 - Request for Additional Information Regarding Inter-Unit Transfer of Spent Fuel Amendment Request," (CAC Nos. MF8991 and MF8992) (ML17219A106)

7) Entergy letter dated August 16, 2017, "Indian Point Nuclear Generating Unit Nos. 2 and 3 - Response to Request for Addltional Information Apb(

~,trc

NL-17-115 Docket Nos. 50-247 and 50-286 Page 2 of 3 Regarding Inter-Unit Transfer of Spent Fuel," (NL-17-102)

(ML17234A402)

Dear Sir or Madam:

a Entergy Nuclear Operations Inc. (Entergy) requested License Amendment [Reference 1] for Indian Point Nuclear Generating Unit Nos. 2 (IP2) and 3 (IP3) Inter-Unit Transfer of Spent Fuel.

On April 11, 2017, the Nuclear Regulatory Commission (NRC) staff identified the need for supplemental information to accept the requested licensing action [Reference 2].

By Reference 3, Entergy provided responses to supplemental information requested by the NRC. By Reference 4, the NRC staff stated their acceptance of the proposed amendment

[Reference 1], along with the supplemental information provided in Reference 3.

By References 5 and 6, the NRC identified a need for additional information in order to complete its review of the proposed license amendment requested. Reference 7 provided the response to the information requested by the NRC in Reference 5.

The attachments to this letter provide the information requested by Reference 6. It should be noted that the response to the RAl-4 necessitated a proposed change to the Table 3.1.1-2 on page 3.1.2-5 of Indian Point Unit 2 and Unit 3 technical specification as reflected in Attachment

2. The information provided herein does not change the conclusion that the proposed changes involve no significant hazards considerations. , Holtec International Report Hl-2084146, "Thermal Hydraulic Analysis of IP3 Shielded Transfer Cask", Rev. 9, and Attachment 4 Holtec International Report Hl-2084109, "Shielding Design Calculation of Transfer Canister for Indian Point 3", Rev. 13, contain information that Holtec consider to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. In addition, Attachment 5 Holtec International Licensing Report Hl-2094289, Rev. 9 contains information that Holtec and Westinghouse consider to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. In accordance with 10 CFR 2.390 and in support of this request for withholding, affidavits executed by Holtec International and Westinghouse are provided in Attachment 7.

There are no new commitments being made in this submittal.

If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-254-6710.

I declare under penalty of perjury that the foregoing is true and correct. Executed on 6c)rokY'"" ,0.,. , 2017.

Sincerely,

~ ;/i AJV/mm /( ~

NL-17-115 Docket Nos. 50-247 and 50-286 Page 3 of 3 Attachments:

1. Reply to NRC Request for Additional Information Regarding Inter-Unit Transfer of Spent Fuel

2. Revised Proposed TS Page Markups

3. Revised Hl-2084146, 'Thermal Hydraulic Analysis of IP3 Shielded Transfer Canister, Rev. 9 (Holtec Proprietary)

4. Revised Hl-2084109, "Shielding Design Calculation of Transfer Canister for Indian Point 3", Rev. 13 (Holtec Proprietary)

5. Revised Hl-2094289, "Licensing Report On The Inter-Unit Transfer of Spent Fuel at The Indian Point Energy Center", Rev. 9 (Holtec and Westinghouse Proprietary)

6. Revised Hl-2094289, "Licensing Report On The Inter-Unit Transfer of Spent Fuel at The Indian Point Energy Center", Rev. 9 (Non-Proprietary Version)

7. Affidavits in Support of Request to Withhold Information cc: Mr. Daniel H. Dorman, Regional Administrator, NRC Region I Mr. William Burton, NRC Senior Project Manager, Division of License Renewal Mr. Richard V. Guzman, NRR Senior Project Manager Ms. Bridget Frymire, New York State Department of Public Service Ms. Alicia Barton, President and CEO NYSERDA NRC Resident Inspector's Office

ATTACHMENT 1 to NL-17-115 REPLY TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE LICENSE AMENDMENT REQUEST FOR INTER-UNIT TRANSFER OF SPENT FUEL ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

NL-17-115 Attachment 1 Page1of17 REPLY TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE LICENSE AMENDMENT REQUEST FOR INTER-UNIT TRANSFER OF SPENT FUEL RAl-1 (CSTB-Thermal)

Request Clarify how the Appendix C TS LCO 3.1.2.b assures the shielded transfer canister (STC) total heat load is not exceeded. LCO 3.1.2.b as shown in Attachments 2 and 4 of the LAR specifies the fuel assembly maximum decay heat. However, it is not clear how the STC total heat load is met when only Type 2 fuel is loaded. Additional assurance may be provided by also specifying the STC total decay heat in LCO 3.1.2.b, as it is done in LCO 3.1.2.a.3.

Regulatory Basis: This information is needed to determine compliance with Title 10 of the Code of Federal Regulations (10 CFR) Sections 72.122 and 72.128.

Response to RAl-1 It is correct that LCO 3.1.2.b in Appendix C of the Tech Spec specifies only the fuel assembly maximum decay heat and not the STC total heat. This is because LCO 3.1.2.b restricts fuel assemblies to only eight basket cells i.e. 5, 6,7,8,9, 10, 11, 12. For a fuel assembly maximum decay heat of 1.2 kW, this ensures that the STC total heat load limit remains below 9.6kW.

If only type 2 fuel assemblies are loaded in the STC, LCO 3.1.2.a is applicable which already specifies both per assembly decay heat and total STC decay heat limits.

RAl-2 (CSTB-Thermal)

Request Explain how the per-cell decay heat and total per-region decay heat would meet the STC total decay heat requirement specified in the LAR. Table 3.2: Loading Scenarios Adopted for Thermal Evaluations of Holtec Report Hl-2084146 in Enclosure 3 of your LAR provides a representative heat load pattern used to perform sensitivity analysis to show the effect on predicted temperatures and pressures. However, it is not clear from the values shown on this table how the STC total decay heat requirements are met. For example, adding the total heat load for Regions 1 and 2 of Loading Scenario 2 would result in a value that would exceed the specified limit for the STC.

Regulatory Basis: This information is needed to determine compliance with 10 CFR 72.122 and 10 CFR 72.128.

NL-17-115 Attachment 1 Page 2 of 17 Response to RAl-2 1

There is a typographical error in Table 3.2 of Holtec report Hl-2084146. The maxim um heat load per cell in the outer region (8 cells) under Loading Scenario 2 is incorrectly stated as 1.2kW instead of 0.6kW. Table 3.2 has been amended to resonate this. The total decay heat in the inner and outer regions is 4.8kW each which sums up to a total heat load limit of 9.6kW, equal to the STC total heat load limit.

RAl-3 (CSRB-Shielding/Radiation Protection)

NRC Request Confirm the following:

Request

a. The dose rate and dose information for the Independent Spent Fuel Storage Installation (ISFSI) in the safety analysis report (SAR) analyses continue to bound the dose contributions of the ISFSI, modifying the SAR analyses as needed. More storage casks have likely been added to the ISFSI since the original amendment to add the STC operations, which could increase the ISFSI contributions to doses. This, in turn, could affect the SAR dose analyses.

b. The activity values shown in Table 7.2.9 are the total burnable poison rod assembly (BPRA) activities which are used in the shielding analysis. If they are in-core activities only, also provide the total BPRA activities and ensure the analyses use the total BPRA activities.

c. Regarding the purpose and use of Appendix G of the shielding calculation package, this appendix was not part of the calculation package submitted with the previous amendment application dated July 8, 2009 (ADAMS Accession No. ML091940176) for the STC operations. Its purpose and use in terms of the shielding analysis is not clear and should be explained.

Regulatory Basis: This information is needed to confirm compliance with 10 CFR 20.1101 (b) and 10 CFR 20.1301(a), (b), and (e) and the intent of 10 CFR 72.104, 10 CFR 72.106(b), and 10 CFR 72.126(a).

Response to RAl-3a We confirm that the reported ISFSI dose information are bounding. ISFSI dose rates taking into account recent cask additions are updated in shielding report Hl-2084109R13 Tables 35-39, and in licensing report Hl-2094289, Tables 7.4.16-7.4.20.

Table 7.4.7, Table 7.4.8, Table 7.4.14, and Table 7.4.15 reference Table 3 of Reference [L.J]

which provides a bounding ISFSI dose rate for a fully loaded 78 cask ISFSI, so no update is required in these tables.

NL-17-115 Attachment 1 Page 3 of 17 Response to RAl-3b We confirm that Table 7.2.9 in fact provides total BPRA activities, not just the portion in the active core region.

With respect to BPRA activities used for the new loading plans 7 through 12, please see also responses to RAls 4 and 5.

Response to RAl-3c Appendix G documents the dose rate calculations at various distances from the loaded STC for ALARA purposes during STC operations. There are no specific acceptance criteria for this analysis, and the results are not used to show compliance with any dose rate requirement in the application. A statement is added at the beginning of this appendix for clarification.

RAl-4 (CSRB-Shielding/Radiation Protection)

Request Confirm the Indian Point 2 and Indian Point 3 TSs in Appendix C for allowable minimum cooling time and maximum burnup for BPRAs will remain unchanged for the proposed new STC loading patterns.

The shielding analysis in the SAR and the supporting shielding calculation package assumes that the cooling time for the BPRAs will be the same as the fuel assembly in which the BPRA is loaded. The analysis also uses burnups of 40 gigawatt-days per metric ton of uranium (GWd/MTU) for those BPRAs in fuel assemblies with burnups not exceeding 40 GWd/MTU and 60 GWd/MTU for those BPRAs in fuel assemblies with burnups in excess of 40 GWd/MTU. It appears that these parameters are intended to specify the allowable BPRA contents for the proposed new STC loading patterns. However, the TSs related to BPRAs (see Table 3.1.2-2 in Appendix C of the licenses) have not been changed and do not allow for loading of BPRAs with burn up and cooling times that fit the above stated descriptions for three of the loading patterns. In particular, the BPRAs loaded into STC basket' cells 1 through 4 of loading patterns 8, 11, and 12 are limited to 50 GWd/MTU, 20 GWd/MTU, and 40 GWd/MTU, respectively. Even for loading patterns 7 and 10, BPRAs in the same four basket cells would be limited to 50 GWd/MTU, versus the 60 GWd/MTU considered per the SAR descriptions. A BPRA with a burnup of 60 GWd/MTU must have a minimum cooling time of not less than 11 years.

I Additionally, per the parameters used in the shielding analysis, BPRAs in cells 1 through 4 of loading pattern 11 could have a burnup up to 60 GWd/MTU. This would appear to be unacceptable since the maximum cobalt-60 activity for such a BPRA with only 6 years of decay would exceed the maximum activity level previously analyzed and used in the shielding and radiation protection analyses. Any proposed changes to the TSs should be supported or bounded by the analysis in the application and clearly confined to the proposed new loading patterns.

Regulatory Basis: This information is needed to confirm compliance with 10 CFR 20.1101 (b) and 10 CFR 20.1301 (a), (b), and (e) and the intent of 10 CFR 72.104, 10 CFR 72.106(b), and 10 CFR 72.126(a).

NL-17-115 Attachment 1 Page 4of17 Response to RAl-4 The limiting burnup and cooling times in the TS are only applicable to patterns 1 through 6. A note is added to the Table 3.1.2-2 to clarify this, and specify the limits for patterns 7 through 12.

The technical specification for both Unit 2 and Unit 3 has been revised to reflect this clarification.

RAl-5 (CSRB-Shielding/Radiation Protection)

Request Clarify the cobalt impurity levels used in the modified analysis for:

a. The lnconel components in the spent fuel assembly's hardware, and

b. The steel and lnconel components of the non-fuel hardware to be loaded with the spent fuel assemblies.

The modified analysis describes a change to the cobalt impurity level in the steel components of the fuel assembly hardware. However, it is not clear if the analysis changed the impurity level for lnconel components of the assembly hardware. Also, it is not clear that the analysis uses the same impurity level for steel and lnconel components of the non-fuel hardware, such as BPRAs. The SAR describes crediting different decay times for the new BPRA analysis for the difference in BPRA cobalt-60 activities. However, the decay times in Table 7.2.9 overlap with decay times in Table 7.2.8 for which the cobalt-60 activity was previously evaluated to be 895 curies. Thus, the decay time does not account for the difference in cobalt-60 activity at the decay times listed in Table 7.2.9, since the activity is smaller at shorter decay times than was previously analyzed. While the SAR continues to state that the cobalt in the non-fuel hardware is taken to be 1.2 grams to kilograms (g/kg) in steel and 4.7 g/kg in lnconel, the analysis information does not appear to be consistent with this statement. Note that the differences in activity, level for BPRAs are not consistent with a 0.5 g/kg impurity level either.

Regulatory Basis: This information is needed to confirm compliance with 10 CFR 20.1101 (b) and 20.1301 (a}, (b}, and (e) and the intent of 10 CFR 72.104, 10 CFR 72.106(b), and 10 CFR 72.126(a).

Response to RAl-5a

(

The cobalt impurity level for lnconel in the spent fuel assembly's hardware is clarified in Table 7.2.10 of Hl-2094289 R9 and in Table H-3 of Hl-2084109 R13. Note 2 of Table 7.1.1 is corrected to refer to Table 7 .2.10. No change was made to the cobalt impurity level in the spent fuel assembly's hardware, i.e. the same value of 4. 7 g/kg is used for lnconel components in the spent fuel assembly's hardware. The third paragraph of Sub-section 7.2.2 of Hl-2094289R9 also provides additional explanation.

Response to RAl-5b While the cobalt impurity level for lnconel in the non-fuel hardware remained unchanged, the cobalt impurity level for stainless steel in BPRAs was changed from 1.2 g/kg to 0.8 g/kg for patterns 7 through 12, and corresponding sections of Hl-2094289 R9 are updated to reflect this.

0.8 g/kg is consistent with the evaluations used in theHl-STORM 100 (Reference [4] in HI-

NL-17-115 Attachment 1 Page 5 of 17 2084109 R13), and supported by the updated comparison between measured and calculated dose rates presented in Appendix I of Hl-2084109 R 13. All cobalt impurity levels of the non-fuel hardware (BPRA, RCCA, and TPD) are clarified in Table 7.2.11 of Hl-2094289 R9 and in Table H-4 of Hl-2084109 R13.

RAl-6 (CSRB-Shielding/Radiation Protection)

Request Modify the analyses to address the uncertainties associated with calculations of source terms for high burnup fuel.

The licensee uses an older version of SCALE (version 4.3) and a module of that code, SAS2H, which is no longer supported by the code developer (Oak Ridge National Laboratory). The NRC staff accepted the use of this code version and code module in the previous analyses for the STC and the proposed contents, which included a limited amount of high burnup spent

• fuel. The NRC staff's safety evaluation forthat license amendment dated July 13, 2012 (ADAMS Accesssion No. ML121230011), however, indicates that it accepted the analysis with code version and code module, in part, because the burnups did not extend significantly into the high burnup regime. Also, high burnup fuel was limited to the inner STC basket locations of only two loading patterns. Burn ups for the proposed additional loading patterns extend further into the high burn up regime, and high burnup fuel may be loaded into inner or outer STC basket locations for multiple loading patterns. Thus, the licensee should address the uncertainties in the analyses for calculating high burnup fuel source terms with the code version and code mo9ule used in the analyses. One possible approach would be to evaluate the uncertainties and adjust the analyses in a manner similar to what has been done for the certified HI-STORM 100 dry storage system. Uncertainties in terms of radiation source terms and decay heats should also be addressed.

Regulatorv Basis: This information is needed to confirm compliance with 10 CFR 20.1101 (b) and 20.1301(a), (b), and (e) and the intent of 10 CFR 72.104, 10 CFR 72.106(b), and 10 CFR 72.126(8).

Response to RAl-6 For the HI-STORM 100, the approach to consider source term uncertainties is based on an uncertainty developed for the heat load calculation with a value of 5%. This then rolls into corresponding burnup limitations through the application of the method in Section 5.2.5.3 of the HI-STORM 100 FSAR. The situation for the STC is slightly different, whereas both burnup and heat load limits are specified explicitly, not through any equations. However, the total maximum theoretical heat load of the cask consistent with the explicit burnup, enrichment and cooling time limits for the assemblies, ranges between 10. 75 and 12.18 kW for the new patterns with HBF assemblies, based on the sum of decay heats arising from individual assembly burnup and cooling time combinations. However, the explicit thermal limit of 9.621 kW prevents reaching those limits. It is therefore not permissible to load every assembly in the cask under those new loading patterns up to the explicitly stated burnup, enrichment and cooling time limits. The corresponding level of heat load margin for the entire cask is between about 10% (1-9.621/10.75) and 20% (1-9.621/12.18). This is significantly larger than the 5% utilized in the approach used for the HI-STORM 100, and while the value may be smaller for an individual assembly, the overall effect would be that the margin available to cover any uncertainty in the

NL-17-115 Attachment 1 Page 6 of 17 source terms of the STC is expected to be at least as large, if not larger than that for the HI-STORM 100. Additionally, note that, as for the HI-STORM 100, the source term calculations are performed using a single continuous cycle, a conservative approach that provides additional margin. Based on all this, we consider that an additional consideration of source term uncertainty or adjustment of results is not necessary.

RAl-7 CCSRB-Shielding/Radiation Protection)

Request Justify the use of proposed contents burnup, enrichment, and cooling times that result in decay heats that exceed the proposed decay heat limits, adjusting the proposed burnup, enrichment, and cooling time specifications as needed.

Based on Table 7.1.1, specifications for fuel assemblies in the inner four STC basket locations in loading pattern 12 result in assembly decay heats that exceed the allowed decay heat limits for those basket locations. The proposed burnup, enrichment, and cooling time specifications should also ensure the decay heat limit is met, considering the contribution from non-fuel hardware as well as the uncertainties in calculations of decay heat and radiation source terms for high burnup fuel (see preceding request for additional information (RAI)). This concern should similarly be addressed in terms of the limit for the total STC decay heat load since all of the new proposed loading patterns result in decay heats that exceed this limit. An alternative approach would be for the licensee to modify the operations descriptions in Chapter 10 of the SAR to ensure that verification of acceptability of assemblies and non-fuel hardware for loading in the STC includes verification that the individual decay heats and the cumulative basket decay heat meet their respective limits. In other words, the verification of the acceptability of assemblies, as described in Chapter 10 of the SAR, should include more than just verification of the assembly burnup, enrichment, and cooling time. The operations should include verification of decay heat of each assembly as well as the entire STC.

Regulatory Basis: This information is needed to confirm compliance with 10 CFR 20.1101 (b) and 20.1301(a), (b), and (e) and the intent of 10 CFR 72.104, 10 CFR 72.106(b), and 10 CFR 72.126(a).

Response to RAl-7 The regionalized burnup, initial enrichment, and cooling time combinations is the shielding discipline's imposed limits on the spent nuclear fuel. The thermal discipline has a separate limit for decay heat (per cell and per basket). Both the thermal and shielding limits for spent fuel assemblies in the Technical specifications must be met for STC loadings. The shielding analysis conservatively uses burnup, initial enrichment, and cooling time combinations that yield a calculated decay heat that exceeds the decay heat limit in Table 5.0.1 to allow for greater loading flexibility. The Technical Specifications, Appendix C, Limiting Condition of Operation (LCO) 3.1.2 includes verification of decay heat of each assembly as well as the entire STC.

Chapter 10 sub-section 10.2.2 step "1.a" ensures "The fuel assemblies and any non-fuel hardware intended for transfer will be characterized to ensure compliance with Appendix C of the TS." Thus, the burnup, enrichment, and cooling time limits for individual fuel assemblies would, by themselves, seem to permit loading an STC beyond its thermal limit. However, the individual limits on decay heat from the assemblies and non-fuel hardware in the TS ensure the thermal limits for each cell location and the STC overall are not exceeded.

I

NL-17-115 Attachment 1 Page 7of17 RAl-8 (CSRB-Shielding/Radiation Protection)

Request Provide information to demonstrate that the current HI-TRAC dose rates are bounding for the additional loading configurations for both normal conditions and accident conditions.

Based on information in the SAR and shielding calculation package (e.g., Tables 23 through 29 of the calculation package), neutron dose rates are a significant or dominant component of the total dose rates for the HI-TRAC for both normal and accident conditions. Additionally, with higher burnups, such as those in the proposed loading configurations 7 through 12 (particularly those with high burnup fuel in the outer basket locations), the neutron source term increases significantly versus the source terms in the configurations currently identified as yielding bounding dose rates. Thus, it would appear that the HI-TRAC dose rates should be updated to address the new loading patterns and the evaluations versus the regulatory limits revised. Sample input files may also be useful in responding to this question.

Regulatory Basis: This information is needed to confirm compliance with 10 CFR 20.1101 (b) and 20.1301 (a), (b), and (e) and the intent of 10 CFR 72.104, 10 CFR 72.106(b), and 10 CFR 72.126(a).

Response to RAl-8 After performing dose rate calculations for loading patterns 7-12 with the STC in the HI-TRAC it was found that the neutron dose rate and total dose rate for loading pattern 8 in some cases is bounding. Previously loading patterns 3 and 4 were bounding for the bare STC and for loading patterns 1-12 they remain bounding for the bare STC. Previously loading patterns 3 ~nd 4 were bounding for the STC in the HI-TRAC, but now loading patterns 4 and 8 are bounding for the STC in the HI-TRAC for loading patterns 1-12. The following changes were made to update tables:

• Tables for the HI-TRAC normal condition have been added to Appendix A for loading patterns 7-12.

• Tables 23, 25, 27, and 29 are updated as Loading Pattern 8 has become bounding loading pattern for neutron dose rates.

• Tables in Appendix A for Accident conditions 1, 2, and 3 have replaced Loading Pattern 3 with Loading Pattern 8.

• Tables 38 and 39 have been updated for HI-TRAC Accident cases 2 and 3 (Loading Pattern 8) respectively to show compliance with 10CFR72.106.

Note that no additional radiation transport calculations were performed for this, all results are from existing calculational models, by just changing source specifications. Hence no new input files were generated.

NL-17-115 Attachment 1 Page 8of17 RAl-9 (CSRB-Shielding/Radiation Protection)

Request Confirm that dose rates from BPRAs are calculated correctly for the proposed loading configurations, revising the calculations as needed.

Based on the information in the SAR and shielding calculation package, the source terms for BPRAs for all the new proposed loading patterns should be for exposures of 60 GWd/MTU since all of the proposed loading pattern limits are for fuel with burn ups exceeding 40 GWd/MTU. Thus, the dose rates due to BPRAs for the new loading patterns should vary from the BPRA dose rates in the design basis calculations in proportion to the ratio of the cobalt-60 activities determined in Table 7.2.9 of the SAR versus the design basis activity of 895 curies. Based on the information in Appendix A of the calculation package and Table 7.4.2 of the SAR, the variation in BPRA dose rates at 1 meter from the STC top surface is consistent with this expected change. However, the BPRA dose rates at 1 meter from the radial and bottom surfaces are not consistent with this expectation; they are substantially below values that would be consistent with this expectation (about 1% or less of the expected values). Thus, it appears there is an error in the analyses for BPRA dose rates, at least for the noted locations and possibly for others. Sample input files for the analyses for the previously approved STC amendment and sample input files for the proposed, modified analyses may be helpful in responding to this question.

Regulatory Basis: This information is needed to confirm compliance with 10 CFR 20.1101 (b) and 20.1301(a), (b), and(e) and the intent of 10 CFR 72.104, 10 CFR 72.106(b), and 10 CFR 72.126(a).

Response to RAl-9 As discussed in the response to RAI 4, the calculations for patterns 7-12 use a lower assumed cobalt content for the BPRAs, which in turn results in lower dose rates.

All new loading patterns assume a BPRA burnup of 60 GWD/MTU as shown in Table 7.2.9 of\

Hl-2094289 R8 and Table H.2 of Hl-2084109 R12 and are confirmed to be calculated correctly.

We recognized that the presentation of results in Appendix A was partially inconsistent and could lead to incorrect conclusions. Therefore, Appendix A STC calculations were reformatted so that the results for Total with BPRAs, Total with CRAs, Total with TPDs, and Total (without non-fuel hardware) are combined on a single page on pages A-16 through A-21 in HI- 2084109R13.

With this, the expected ratio of BPRA contributions can be more easily confirmed. For example, for two patterns with comparable fuel in the outer region (Pattern 4 and Pattern 11, Pages A-12 1

and A-20), BPRA dose rates (difference between 'Total" and 'Total with BPRA") for the side at 1 m (Tally 4001, Segment 2), are about 570 and 240 mrem/hr, respectively. The ratio between these compare very well with the activity ratio, using the value for 14 years cooling time for pattern 11 (about 385 Ci) and 895 Ci for pattern 4.

Note that no additional radiation transport calculations were performed for this, all results are from existing calculational models, by just changing source specifications. Hence no new input

NL-17-115 Attachment 1 Page 9 of 17 files were generated.

RAl-10 (CS RB-Shielding/Radiation Protection)

Request Provide information to demonstrate that the proposed change to the analysis for control rod assemblies (RCCAs) is adequately bounding for actual operations of RCCAs that are not typical, revising the dose rate calculations as necessary.

The licensee proposed to modify the source term calculation for RCCAs based on typical operating parameters for RCCAs at IP3. However, the analysis should be bounding for RCCAs which may have experienced operations that are not consistent with typical operations at the plant (e.g., operated with some insertion into the core for one or more periods of time or use as regulating rods), if any. Since the information in the SAR only discusses typical RCCA operations, it is not clear how the proposed analysis change considers RCCAs that have experienced non-typical operations. ,

Regulatory Basis: This information is needed to confirm compliance with 10 CFR 20.1101 (b) and 20.1301(a), (b), and (e) and the intent of 10 CFR 72.104, 10 CFR 72.106(b), and 10 CFR 72.126(a).

Response to RAl-10 Indian Point operates with all control rods fully withdrawn during full power operation. The vast majority (>90%) of RCCAs at IPEC are fully withdrawn during all power operations, meaning that the lesJ activated RCCA Configuration 2 would be most appropriate for the vast majority of IPEC RCCAs. During reduced power operations, which take place during start-up and shutdown, a small subset of control rods may be partially inserted (less than 10% insertion). The periods of reduced power operations take significantly less than 10% of the total operating cycle time at IPEC. Therefore, RCCA Configuration 3, which assumes 25% of the activation of RCCAs that have been inserted 10% into the core for the all power operations, bounds all IPEC RCCAs (See Table 7.2.7) that meet RCCA Tech Spec Requirements (Table 7.2.8).

Conservatively, RCCA Configuration 3 is assumed for all IPEC RCCAs for loading patterns 7-12.

Additionally, please note that the comparison between calculated and measured dose rates documented in Appendix I of Hl-2084198 R 13 still show substantial margin at the bottom of the STC where the contribution from RCCAs would be highest, even though the contributions from RCCAs were excluded from the calculations.

RAl-11 (CSRB-Shielding/Radiation Protection)

Request Modify the evaluation of measured dose rates vs. calculated dose rates for the two STC loads described in Appendix I of the shielding calculation package to address the following:

NL-17-115 Attachment 1 Page 10of17

a. Calculate dose rates using burnup, enrichment and cooling time values for the assemblies that are closest to the measurement location for which the dose rate comparison is made.

b. Compare dose rates for locations which are adequately defined to know where on the STC the measurement was made and where the dose rates should be calculated for accurate comparison.

c. Calculate dose rates, for burnups and cooling times of the non-fuel hardware that are consistent with the burnups and cooling times of the hardware in the loaded STC.

d. Clarify the descriptions of the measurements to (1) identify the assemblies that were closest to the measurement locations, (2) identify the location of the measurement at the STC base (e.g., at the center of the base or the relative location versus the center or the inner region of the STC basket), (3) confirm the measurements are surface measurements (vs. at distance from the STC surfaces), and (4) provide the (estimated) uncertainties in the measurements.

The comparisons in Appendix I of the calculation package are used to demonstrate the degree of conservatism that remains even with the changes to the source terms for the spent fuel assemblies and non-fuel hardware specifications for the proposed loading patterns. However, it is not clear that the current evaluation is adequate for that purpose. Based on the information provided in the appendix, it is not clear that the calculations are for the same locations on the STC as the measurements. At least in one instance, it is not clear where the measurement was taken that is used in the comparison (see Table 1.2.B). Significant variations in dose rates are expected depending upon the measurements' locations.

Also, the conservative nature of the analysis method should be demonstrated with calculations that use the actual burnups, enrichments, and cooling times for the contents (both the spent fuel and the non-fuel hardware). Differences in burnup, enrichment and cooling time of assemblies versus the region average used in the calculation can influence the determination of the degree of conservatism that may or may not be present in the analysis method. This may be particularly true for the assemblies in the outer STC basket region, especially for STC #1. Thus, the licensee should use the burn ups, enrichments, and cooling times of the assemblies that are closest to the measurement locations to determine dose rates for comparison with measured dose rates. Alternatively, the licensee may represent the assemblies in the region using the assembly that was loaded in that region of the STC and has burnup, enrichment, and cooling time parameters that result in the weakest radiation source term (and thus the lowest dose rates). In addition, for the non-fuel hardware, the source term should be derived from burnups and cooling times of the loaded hardware and not the design basis source terms. If design basis non-fuel hardware were loaded in the STC (as allowed by the TSs), then use of the design-basis hardware source term would be appropriate for this analysis.

Regulatory Basis: This information is needed to confirm compliance with 10 CFR 20.1101 (b) and 20.1301 (a), (b), and (e) and the intent of 10 CFR 72.104, 10 CFR 72.106(b), and 10 CFR 72.126(a).

NL-17-115 Attachment 1 Page 11of17 Response to RAl-11 In order to enhance the comparison between measured and calculated dose rates, and to demonstrate the degree of conservatism that remains, even with the changes to the source terms for the spent fuel assemblies and non-fuel hardware specifications for the proposed loading patterns, we expanded the evaluations in the following areas:

• Previously, average burnup, enrichment and cooling time values were used for each of the two regions in the basket. The revised calculations allow to allocate a burnup, enrichment and cooling time individually to each assembly in the outer region, and each insert.

• Holtec typically uses pre-calculated source terms, generated for burnups, enrichments and cooling times on prescribed increments (e.g. multiples of 5000 MWd/mtU for burnup). The source terms closest to the actual fuel specifications are then used. To indicate the impact of the difference between actual fuel and the parameters used in the source term calculations, upper bound and lower bound values are used.

• The documentation on the measurements is limited, since the measurements were performed as part of the radiation protection activities, i.e. to simply identify high dose rate areas, hence the measurement location had not been selected or documented with the level of detail in mind needed for a rigorous comparisons with calculations. To overcome and address possible limitation from this, the following expansions were made:

o Measurements were recorded for both surface and at a distance of 30 cm from the surface. The initial comparison only utilized the surface dose rates for comparison. The expanded comparison also utilizes the measurements at 30 cm. At that distance, the measurements present a more integral value, and hence rely much less on the precise knowledge of the measurement location.

o With the consideration of the actual loading configuration of the assemblies, azimuthal variations in calculated dose rates are determined, and the range of the dose rates around the cask is compared to the measured value. This is specifically important since the exact azimuthal locations of the measurements may not be precisely known in all cases. But it also gives an indication of the spread of results over the circumference of the cask, at both the surface and at 30 cm distance from the surface.

• Finally, to assess the impact of the revised assumptions used in the analyses, selected calculational results are presented with the original more conservative assemblies, and the revised assumptions used for the additional patterns presented in this amendment.

The expanded evaluations support the following conclusions:

• *While some ratios between calculated and measured values are slightly higher than before, the comparisons in general confirm the conservative nature of the calculations.

• The added comparison of dose rate values at 30 cm shows in general the same trends as those on the surface.

This addresses the four sub-questions a) through d) in the RAI as follows:

Dose rates are now calculated considering individual burnup, enrichment and cooling times of assemblies, and in various locations. This allows a conservative comparison even when the exact location of the measurement is not clear.

NL-17-115 Attachment 1 Page 12 of 17 a) Dose rates are calculated in various locations. This allows a conservative comparison even when the exact location of the measurement is not clear.

b) Dose rates are now calculated considering individual burnup, enrichment and cooling times of the non-fuel hardware in each location in the outer region. For conservatism, the dose rates from the control rod assemblies in the assemblies in the inner region are neglected in the calculations. '

c) The available descriptions of the measurements, including the measurement uncertainty, are added to the Appendix.

Overall, the expanded comparisons provide additional assurance that the proposed additional loading patterns, with higher burned assemblies on the periphery of the STC will not result in any substantial new challenges from a radiation protection perspective for the upcoming loading campaigns.

RAl-12 (CSRB-Criticality Safety)

Request Clarify which part(s) of the Chapter 4 SAR has been changed and provide a revised SAR, if necessary, to clearly mark the changes and provide an evaluation of the impact of the changes on criticality safety of the STC.

On page S7 of the LAR dated December 14, 2016, the licensee states: "All of Chapter 4 has significant changes due to the change in criticality methodology from Part 50 to Part 71. This is summarized in the preface to chapter 4 responses in the response to the RAI." The licensee further states on page 14 of Attachment 1: "New criticality evaluations for both the STC confirm that operation in accordance with the proposed amendment continues to meet the required subcriticality margins." However, only a very small portion (shown only on page 4-43 and 4-57) of Chapter 4 of the SAR is marked with revision bar. As such, the staff was unable to identify the changes in the SAR and new criticality evaluations; consequently, the staff is unable to evaluate the impact on criticality safety. The licensee is requested to provide a revised SAR in which the changes are clearly marked. The licensee is also requested to add an evaluation of the impact of the changes, item by item, on criticality safety of the STC.

Regulatory Basis: This information is needed to determine if the Indian Point spent fuel inter-unit transfer canister STC design meets the regulatory requirements of 10 CFR 72.124(a) and 10 CFR 72.124(b).

Response to RAl-12 In the current revision, there are no changes in the criticality analysis, and there are no changes to the loading table, and there is no change in the burnup credit taken for fuel assemblies with burnup beyond what was previously analyzed in Revision 6 of the licensing report.

The beginning of the licensing report contains a brief characterization of the changes for ALL revisions of the report. The change on page S4 of S7 relates to a much earlier version of the report which was previously approved and incorporated from Revision 1 to Revision 3.

The changes to Chapter 4 in the current revision of the iicensing report are characterized on Page S2. The changes are contained in Sections 4. 7 and 4.8.

NL-17-115 Attachment 1 Page 13 of 17 The change in Section 4. 7 provided additional justification for the margin available to cover basket manufacturing tolerances. No changes to the methodology were made.

The changes in Section 4.8 are editorial in nature regarding the spent fuel pool. Section 4.8 also added the restriction that IPEC Unit 3 fuel can only be stored in IPEC Unit 2 spent fuel pool rack region 1-2. In addition, Appendix 4.B, which contains Section 4.8 of Revision 6 was added to the current revision for information only. The changes in the current revision are marked with revision bars.

The changes made in Chapter 4 of the current revision do not impact the criticality safety of the STC.

RAl-13 (CSRB-Criticalitv Safety)

Request Revise Table 3.1.2-3, "Allowable STC Loading Configurations" to include the upper enrichment limit and the required minimal burn up for each of the proposed configurations that takes burnup credit.

On page 3.1.2-7 of the proposed new loading table, Table 3.1.2-3, the licensee specifies minimum enrichment and maximum burnup for each configuration. However, the proposed loading table does not include maximum fuel enrichment limit and minimal burnup requirement for each configuration. Because the Shield Transfer Canister (STC) takes credit for fuel burnup, the loading table must include a maximum enrichment and minimal burnup for each configuration and/or location in fuel basket that takes burnup credit. The licensee is requested to revise Table 3.1.2-3 to include the maximum enrichment limit and the required minimal burnup for each configuration.

Regulatory Basis: This information is needed to determine if the Indian Point spent fuel inter-unit transfer canister STC design meets the regulatory requirements of 10 CFR 72.124(a).

Response to RAl-13 Please note that in the current revision, there are no changes in the criticality analysis.

Therefore, there are no changes to Table 3.1.2-3 because there are no changes to the burnup credit analysis. Technical Specification Table 3.1.2-3 lists the burnup and enrichment limits for each configuration. No additional information is provided since no changes were made in the current revision. The current revision is entirely consistent with Revision 6, with the exception of some minor editorial items.

RAl-14 (CSRB-Criticality Safety)

Request Clarify if damaged fuel is intended to be transferred by the transfer cask. If so, provide a criticality safety analyses consistent with the fuel conditions and intended configurations. One of the proposed changes to the TSs is to change LCO 3.1.2.c from "Only INTACT FUEL ASSEMBLIES with initial enrichment~ 3.2 and s 4.4 w/o U-235 and discharged prior to IP3 Cycle 12 shall be placed in the STC basket" to the various configurations as defined in Table

NL-17-115 Attachment 1 Page 14 of 17 3.1.2-3 of Attachment 2 to the revised SAR. However, there is no restriction in the configurations presented in Table 3.1.2-3 on whether the fuel shall be intact or not. As such, with the proposed changes, the restriction that all fuel to be transferred by the fuel transfer cask shall be intact would be eliminated from the TSs. However, in Table 10.0.1 on page 10-2 of the SAR, the licensee states: "Failed fuel assemblies (i.e. assemblies that are not intact) and/or damaged fuel are not permitted for transfer in the STC." In addition, the criticality safety analyses for the STC does not include any analyses for the transfer cask containing damaged fuel. The licensee is requested to clarify whether damaged fuel is the intended content to be transferred from IP3 spent fuel pool to the IP2 spent fuel pool using the STC. If so, provide criticality safety analyses consistent with the fuel conditions and intended configurations.

Regulatory Basis: This information is needed to determine if the Indian Point spent fuel inter-unit transfer canister STC design meets the regulatory requirements of 10 CFR 72.124(a).

Response to RAl-14 Damaged fuel is not intended to be transferred by the transfer cask atJhis time. Please see that LCO 3.1.2c has been restored (per response to first set of RAls from NRC) and reworded to include the condition that the initial enrichment is less than or equal to 4.4 wt% U-235. The conditions placed on the enrichment and burnup of the fuel assemblies allowed for storage are described more explicitly in Table 3.1.2-3 of the Technical Specifications for the STC. No changes were made to the criticality analysis in the current revision.

RAl-15 (CSRB-Criticality Safety)

Request Clarify the definition for the term "simulated actinide compositions" and justify the validity of this approach for benchmarking analyses.

On page 4.A-1 of the SAR, the licensee states:

Since then, an expanded set of critical experiment has been analyzed [L.O], that includes critical experiments with simulated actinide compositions of spent fuel. This results in a bias of -0.0013 with an uncertainty 0.0086 (95/95), i.e. values ,that are very similar to those used in the HI-STAR 100. Those values were used in all calculation in the Tables in the main part of Chapter 4 of this report where a maximum kett is reported.

However, the term "simulated actinide compositions of spent fuel" was not defined in the SAR. The licensee is requested to clarify the definition of this term and provide justification for the use of "simulated spent fuel compositions" in code benchmarking analyses for burnup credit application. If the term means data that is not obtained from radiochemical assay (RCA) samples, the licensee is requested to provide information on the extent of the use of data. If a significant amount of surrogate data is used in the code benchmarking analyses, the licensee is requested to provide justification for the validity of the approach .

. Regulatory Basis: This information i~ needed to determine if the Indian Point spent fuel inter-unit transfer canister STC design meets the regulatory requirements of 10 CFR 72.124(a).

NL-17-115 Attachment 1 Page 15of17 Response to RAl-15 The MCNP benchmark report referenced on page 4.A-1 of the SAR as "[L.O]" documents the criticality benchmarking which include the MOX configurations from the IHECSBE (International Handbook of Evaluated Criticality Safety Benchmark Experiments) experiments and the HTC (Haut Taux de Combustion) experiments. The benchmarking is performed consistent with the guidance in NUREG/CR-7109. The set of experiments includes critical experiments with simulated actinide compositions. The simulated actinide compositions used in the experiments represent the composition of the fuel with the assumed burnup. The "simulated actinide compositions" on Page 4.A-1 refer to the MOX configurations used in the IHECSBE experiments and the HTC experiments. Note that the benchmarking has not been changed for the current revision of the SAR. :1 RAl-16 (CSRB-Criticality Safety)

Request Clarify if fresh fuel critical experiments are used for the criticality code benchmarking analyses for STC configurations that take burnup credit. If so, provide justification for the validity of this approach or revise the benchmarking analyses with appropriate critical experiments per the recommendation provided in NUREG/CR-7109, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (kett) Predictions" (ADAMS Accession No. ML12116A128).

On page 4.A-1 of the SAR, the licensee states: "HI-STAR 100 initially use critical experiments with fresh and mixed uranium and plutonium oxides (MOX) fuel that results in a bias of -0.0004 with an uncertainty of 0.0083 (95/95). This is discussed in the appropriate section of this appendix, but was only used for the initial scoping calculation for this project." The licensee further states: "Since then, an expanded set of critical experiment has been analyzed [L.O], that includes critical experiments with simulated actinide compositions of spent fuel." As such, it is not clear if fresh fuel critical experiments were used in the final code benchmarking analyses. Based on the latest research results published in NUREG/CR-7109, fresh fuel critical experiments are not suitable for code benchmarking for burnup credit applications. Specifically, NU REG/CR 7109 states:

Based on the analyses performed for this report, it is affirmed that criticality analysts should continue to validate BUC [burnup credit] criticality safety evaluations to the extent possible, with the best available critical experiment data. MOX configurations from the IHECSBE [International Handbook of Evaluated Criticality Safety Benchmark Experiments]

and the HTC [French Huat Taux De Combustion] experiment configurations, collectively, provide sufficient data for validation of BUC analyses with major actinides and hence should be used for validation. LEU (low-enrichment uranium) critical configurations should not be used in a conventional validation analysis to validate burned fuel systems because they do not include any bias contribution from the plutonium present in burned fuel. The validation statistical analysis should include, bias trending analysis as a function of plutonium content, using a trending variable such as plutonium fraction (i.e., gram of Pu per gram of Pu + U).

NL-17-115 Attachment 1 Page 16 of 17 The licensee is requested to clarify if fresh fuel critical experiments are used for criticality code benchmarking analyses for STC that takes burnup credit. If so, provide justification for the validity of this approach or revise the benchmarking analyses with appropriate critical experiments.

Regulatory Basis: This information is needed to determine if the Indian Point spent fuel inter-unit transfer canister STC design meets the regulatory requirements of 10 CFR 72.124(a).

\

Response to RAl-16 No changes are made in Revision 8 of the licensing report for the burnup credit taken for fuel assemblies. No additional burn up credit is taken for fuel assemblies with burn up beyond previously analyzed conditions. The benchmarking has not been changed for Revision 8 of the licensing report. The final code benchmarking used simulated actinide compositions of spent fuel as described in Appendix 4.A. See discussion in response to RAl-15.

RAl-17 (CSRB-Criticality Safety)

Request Provide criticality safety analyses for these configurations corresponding to the revised loading table for the STC fuel loading configurations that take burn up credit.

The licensee states that it performed criticality safety for the STC using the method from HI-STAR 100 criticality safety analysis. However, the NRC staff was unable to find the analyses for the STC except some general discussion of the applicability of the methodology. Although the approach used for the HI-STAR 100 criticality was accepted by the NRC staff, the licensee is expected to perform specific analyses for the STC with various combinations of enrichment and burnup to support the revised loading table per RAl-13, as appropriate. An MCNP (computer code) input or output file containing the material composition data for the most reactive STC model using burnup credit would suffice.

Regulatory Basis: This information is needed to determine if the Indian Point spent fuel inter-unit transfer canister STC design meets the regulatory requirements of 10 CFR 72.124(a).

Response to RAl-17 There are no changes to the loading table in Revision 8 of the licensing report submitted with the amendment request. The criticality safety analysis to support STC loading was approved in a prior revision. No additional information is provided since no changes were made to the STC fuel loading configurations.

RAl-18 (CSRB-Criticality Safety)

Request Provide the fuel material composition for the most reactive STC configuration that takes burnup credit.

The licensee states that it performed criticality safety for the STC using burnup credit. However, the SAR does not include the spent fuel material composition. The NRC staff requests the

NL-17-115 Attachment 1 Page 17of17 material composition of spent fuel that is used in the MCNP model to verify the criticality safety analysis for the most reactive STC configuration that takes burnup credit. An input file for the model of the STC that exhibits the maximum reactivity would suffice.

Regulatory Basis: This information is needed to determine if the Indian Point spent fuel inter-unit transfer canister STC design meets the regulatory requirements of 10 CFR 72.124(a).

Response to RAI-18 There is no change in the burnup credit taken for fuel assemblies in Revision 8 of the licensing report. The criticality safety analysis for STC loading was approved in a previous revision and no changes regarding fuel material composition or burnup credit are being made in Revision 8.

All input files were sent to support review of previous revisions. No additional information is provided since no change has been made in Revision 8 of the licensing report.

ATTACHMENT 2 to NL-17-115 REVISED PROPOSED TS PAGE MARKUPS ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286 j

STC Loading 3.1.2 Table 3.1.2-2 NON FUEL HARDW ARE(a) Post Irradiation Cooling Times and A llowable Average Bumup Max imum Bumup Post-i1rndiation (MW DIM TU)

Cooling Time (years) BPRAs and TPDs(b)(c) Hafnium Flux WABAs (b . ~ RCCAs Suppressors

~ 6 s 20000 N IA s 630000 s 20000

~ 7 - s 20000 - -

~ 8 s 30000 - - s 30000

~ 9 s 40000 s 30000 - -

~ 10 s 50000 s 40000 - -

~ 11 s 60000 s 45000 - -

~ 12 - s 50000 - -

~ 13 - s 60000 - -

~ 14 - - - -

~ 15 - s 90000 - -

~ 16 - s 630000 - -

~ 20 - - - -

Allowed Up to twelve Up to twelve Up to four (4) Up to four (4)

Quantity and (12) per (12) per per transfer in per transfer in Location transfer in any transfer in Cells 1, 2, 3, Cells 1, 2, 3, location anv location and/or4 and/or 4 (a) NON-FUEL HARDWARE bumup and cooling time limits are not applicable to Instrument Tube Tie Rods (ITTRs), since they are installed post-irradiation. NSAs are not authorized for loading in the STC.

(b) Linear interpolation between points is only permitted for BPRAs, W ABAs , and TPDs, with the exception that interpolation is not permitted for TPDs with bumups greater than 90 GWd/MTU and cooling times greater than 15 years .

(c) N IA means not authorized for loading at this cooling time.

(d) ~urnup and Cooling time l~mitt:s in thi oohmµi are only applicable to Loading Patt~rns J-6 m Table 3.1.2-3. For Loadmg Patterns 7-12 m Table 3.l.2-3, the bumup ancf coohng time limits for a BPRA are the same as tho e for the fuel assemlbly they are located in.

INDIAN POINT 2 3.1.2-5 Amendment U&TBD

STC Loading 3.1.2 Table 3.1.2-2 NON FUEL HARDW ARE(a) Post Irradiation Cooling Times and Allowable Average Bumup Max imum Burnup Post-iITadiation (MW DIM TU Cooling Time (years) BPRAs and TPDs(b)(c) Hafnium Flux WABAs(b .< RCCAs Suppressors

~6  :'.S 20000 NIA  :'.S 630000  :'.S 20000

~7 -  :'.S 20000 - -

~ 8  :'.S 30000 - -  :'.S 30000

~ 9  :'.S 40000  :'.S 30000 - -

~ 10  :'.S 50000  :'.S 40000 - -

~ 11  :'.S 60000  :'.S 45000 - -

~ 12 -  :'.S 50000 - -

~ 13 -  :'.S 60000 - -

~ 14 - - - -

~ 15 -  :'.S 90000 - -

~ 16 -  :'.S 630000 - -

~ 20 - - - -

Allowed Up to twelve Up to twelve Up to four (4) Up to four (4)

Quantity and (12) per (12) per per transfer in per transfer in Location transfer in any transfer in Cells 1, 2, 3, Cells 1, 2, 3, location anv location andlor4 and/or 4 (a) NON-FUEL HARDWARE burnup and cooling time limits are not applicable to Instrument Tube Tie Rods (ITTRs), since they are installed post-irradiation. NSAs are not authorized for loading in the STC.

(b) Linear interpolation between points is only permitted for BPRAs, W ABAs, and TPDs, with the exception that interpolation is not permitted for TPDs with burnups greater than 90 GW dlMTU and cooling times greater than 15 years.

(c) NIA means not authorized for loading at this cooling time.

(d Burnup and Cooling ltime limits in 'this column are only applicable to Loading Patterns 1-6 in Table 3. 1.2-3. For Loading Patterns 7-12 in Table 3.1.2-3, the burnup ancf cooliing time limits for a BPRA are the same as those for \the fuel assembly they are located in.

INDIAN POINT 3 3.1.2-5 Amendment +/-46TBD

ATTACHMENT 7 to NL-17-115 AFFIDAVITS IN SUPPORT OF REQUEST TO WITHHOLD INFORMATION ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

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(5) The information sought ito be \Vitbheld is being submitted 1to the me in confidence. The information (including that compiled :from many sources} is of a sort customarily held .in confidence by Holtec International,. and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International. No public disclosure bas been made, and it :is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for 2of5

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Attachment 1 NRC Cover Letter and Affidavit for Use by Entergy

Westinghouse Non-Proprietary Class 3

@Westinghouse *westinghouse Electric Company

• 1000 Westinghouse*orive

. Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission . Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 940-8560 11555 Rockville Pike e-mail: greshaja@westinghouse.com 1 Rockville, MD 20852 CAW-17-4645

• October 2, 2017 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Report HI-2094289, "Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at the Indian Point Energy Center," Revision 9 (Proprietary)

The Application for Withholding Proprietary Information from Public Disclosure is submitted by Westinghouse Electric Company LLC ("Westinghouse"), pursuant to the provisions of paragraph (b)(1) of Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence.

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-17-4645 signed by the owner of the proprietary information, Westinghouse. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Entergy Nuclear Operations, Inc.

Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-17-4645 and should be addressed to James A. Gresham; Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.

l~Regulatory Compliance

© 2017 Westinghouse Electric Company LLC. All Rights Reserved.

CAW-17-4645 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

SS COUNTY OF BUTLER:

\_

I, Jmnes A. Greshmn, run authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC ("Westinghouse") and declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

Executed on: ~'~()'-i(. .Ji~t_1 Jmnes A. Gresham, Manager Regulatory Compliance

)

3 CAW-17-4645 (1) I am Manager, Regulatory Compliance, Westinghouse Electric Company LLC ("Westinghouse"),

and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CPR Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

\._

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

4 CAW-17-4645 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

( c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(iii) There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

5 CAW-17-4645 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, is to be received in confidence by the Commission.

(v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best. of our knowledge and belief.

(vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in Report HI-2094289, Revision 9, pages 4-24, 4-33, and 4-36 (Proprietary), for submittal to the Commission, being transmitted by Entergy Nuclear Operations, Inc. letter. The proprietary information as submitted by Westinghouse is that associated with the inter-unit transfer of spent nuclear fuel between Indian Point Unit 2 and 3, and may be used only for that purpose.

l 6 CAW-17-4645 (a) This information is part of that which will enable Westinghous~ to:.

(i) Assist customers in obtaining licensing changes.

(ii) Assist customers in analyzing the spent fuel pool and absorber panels to ensure criticality does not occur.

(b) Further, this information has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of similar information to its customers for the purpose of assisting in obtaining .licensing changes.

(ii) Westinghouse can sell support and defense of spent fuel pool criticality analyses.

(iii) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to b,e performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and non-proprietary v:ersions of a document, furnished to the NRC associated with the inter-unit transfer of spent nuclear fuel between Indian Point Units 2 and 3, and may be used only for that purpose.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (t) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily ho Ids in confidence identified in Sections (4)(ii)(a) through (4)(ii)(t) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1 ).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, pe~it, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Entergy Nuclear Operations, Inc.

Letter for Transmittal to the NRC The following paragraphs should be included in your letter to the NRC Document Control Desk:

Enclosed are:

1. Report HI-2094289, "Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at the Indian Point Energy Center," Revision 9 (Proprietary)

2. Report HI-2094289, "Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at the Indian Point Energy Center," Revision 9 (Non-Proprietary)

Also enclosed are the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-17-4645, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.

As Item 1 contains information proprietary to Westinghouse Electric Company LLC ("Westinghouse"), it is supported by an Affidavit signed by Westinghouse, the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission ("Commission") and addresses with specificity the considerations listed in paragraph (b )(4) of Section 2.3 90 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse Affidavit should reference CAW-17-4645 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.

f' Attachment 3 Pages Containing Westinghouse Proprietary Information Non-Proprietary Version of Report Hl-2094289, Rev. 9

© 2017 Westinghouse Electric Company LLC. All Rights Reserved.

Westinghouse Non-Proprietary Class 3 Table 4.5.1 Fuel Assembly Specification Assembly type 1 2 3 15x15 Vantage 5, Vantage+, Vantage Description 15x15 LOPAR 15x15 LOPAR P+N+, Upgraded Fuel Fuel Rod Data Fuel pellet outside diameter, in. 0.3659 0.3649 Cladding inside diameter, in. 0.3734 Cladding outside diameter, in. 0.422 Cladding material Zr Stack density, g/cc 96.5%TD Fuel Assembly Data Fuel rod array 15x15 Number of fuel rods 204 Fuel rod pitch, in. 0.563 Max. ZrB2 Coating Loading (g 10

[ r,c (116 rods).

B/cm) or None

[ re (148 rods)

Max. ZrB2 Coating Length, in. T 128 None Number of Instrument/Guide 21 Tubes Guide Tube Material Zr Guide Tube inside diameter, in. 0.498 and 0.499 0.512 Guide Tube outside diameter, in. 0.532 and 0.533 0.546 Active fuel Length, in. 144 Axial Blankets Yes No t Not used in the analyses.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 I 4-24 I Rev. 9

Westinghouse Non-Proprietary Class 3 Table 4.5.6 Hafnium Flux Suppressor Rods Parameter Value Max Number of rodlets per assembly 16 Hafnium Rod OD (in.) 0.3810 Poison Material Hf Absorber Length (in.) [ r,c Assembly Bumup while absorber is 6.1 presentt (GWd/MTU) t Not used in the analysis HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 I 4-33 I Rev. 9

Westinghouse Non-Proprietary Class 3 Table 4.5.9 Fuel Tolerances Parameter Tolerance Increased Fuel Density Not applicable, since all calculations use a bounding fuel density.

Increased Fuel Enrichment + o.o5 wt% 235U Fuel Rod Pitch [ r,e Fuel Rod Cladding Outside [ re Diameter Fuel Rod Cladding Inner Diameter [ re Fuel Pellet Outside Diameter [ r*e Guide Tube Outside Diameter [ re Guide Tube Inside Diameter [ re HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 I 4-36 I Rev. 9 J

a,c a

The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

c Its use by a competitor would reduce his expendihrre of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure ofresomces at our expense.

HOLTEC INTERNATIONAL COPYRIGHTED MATERIAL SHADED AREAS DENOTE HOLTEC PROPRIETARY INFORMATION REPORT HI-2094289 I 4-43 I Rev. 9